Published 1980
| Version v1
Report
Open
OTSG modeling for the analysis of the TMI incident
Description
The IRT code is a reactor plant systems code for the analysis of pressurized water reactor (PWR) transients that is being modified by Brookhaven National Laboratory for the Nuclear Regulatory Commission. A significant modification that has recently been implemented is a model for a once-through steam generator (OTSG) to enable simulation of Babcock and Wilcox (B and W) reactor transients. This paper describes the new OTSG modeling and presents the results of an analysis of the initial phase of the TMI incident using this model
Availability note (English)
MF available from INIS under the Report Number; Available from NTIS., PC A02/MF A01.Files
11552401.pdf
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Additional details
Publishing Information
- Imprint Pagination
- 5 p.
- Report number
- BNL-NUREG--27185
Conference
- Title
- American Nuclear Society annual meeting.
- Dates
- 8 - 13 Jun 1980.
- Place
- Las Vegas, NV, USA.
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 11552401
- Subject category
- S61: RADIATION PROTECTION AND DOSIMETRY; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- COMPUTER CALCULATIONS; HYDRAULICS; MATHEMATICAL MODELS; PRESSURE GRADIENTS; PWR TYPE REACTORS; REACTOR ACCIDENTS; REACTOR SAFETY; STEAM GENERATORS; THREE MILE ISLAND-2 REACTOR; TRANSIENTS
- Descriptors DEC
- ACCIDENTS; BOILERS; ENRICHED URANIUM REACTORS; POWER REACTORS; REACTORS; SAFETY; THERMAL REACTORS; VAPOR GENERATORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-800607--23.